Nuclear Engineering and Technology

  • 제33권5호
  • /
  • Pages.468-482
  • /
  • 2001
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지

An Assessment on the Containment Integrity of Korean Standard Nuclear Power Plants Against Direct Containment Heating Loads

  • 발행 : 2001.10.01
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

As a process of Direct Containment Heating (DCH) issue resolution for Korean Standard Nuclear Power Plants (KSNPs), a containment load/strength assessment with two different approaches, the probabilistic and the deterministic, was performed with all plant-specific and phenomena-specific data. In case of the probabilistic approach, the framework developed to support the Zion DCH study, Two-Cell Equilibrium (TCE) coupled with Latin Hypercubic Sampling (LHS), provided a very efficient tool to resolve DCH issue. In case of the deterministic approach, the evaluation methodology using the sophisticated mechanistic computer code, CONTAIN 2.0 was developed, based on findings from DCH-related experiments or analyses. For three bounding scenarios designated as Scenarios V, Va, and VI, the calculation results of TCE/LHS and CONTAIN 2.0 with the conservatism or typical estimation for uncertain parameters, showed that the containment failure resulted from DCH loads was not likely to occur. To verify that these two approaches might be conservative , the containment loads resulting from typical high-pressure accident scenarios (SBO and SBLOCA) for KSNPs were also predicted. The CONTAIN 2.0 calculations with boundary and initial conditions from the MAAP4 predictions, including the sensitivity calculations for DCH phenomenological parameters, have confirmed that the predicted containment pressure and temperature were much below those from these two approaches, and, therefore, DCH issue for KSNPS might be not a problem.

키워드

참고문헌

  1. SECY-93-087, 'Policy Tecnical, and Licensing Issues Pertaining to Ecolutionary and Advanced Ligth-Water Reactor (LWR) Designs,' USNRC, April, 2, (1993)
  2. KURD, ' Korean Utility Requirements Document, Chapter 5: Engineered Safety Systems,' Rev. 0, KEPCO, June, (1998)
  3. M. M. Pilch, et al., 'The Probability of Containment Failure by Direct Containment Heating in Zion,' NUREG/CR-6075, SAND93-1535, Sandia National Laboratories, December, (1994)
  4. M. M. Pilch, et al., 'The Probability of Containment Failure by Direct Containment Heating in Zion,' NUREG/CR-6075 Supplement 1, SAND93-1535, Sandia National Laboratories, December, (1994)
  5. M. M. Pilch, et al., 'The Probability of Containment Failure by Direct Containment Heating in Surry,' NUREG/CR-6109, SAND93-2078, Sandia National Laboratories, May, (1995)
  6. M. M. Pilch, et al., 'Resolution of the Direct Containment Heating Issue for All Weasinghouse Plants with Large Dry Containments or Subatmospheric Containments,' NUREG/CR-6338, SAND93-2381, Sandia National Laboratories, February (1996)
  7. M. M. Pilch, et al., 'Resolution of the Direct Containment Heating Issue for Combustion Engineering Plants and Babcok & Wilcox Plans,' NUREG/CR-6475, SAND97-0067, Sandia National Laboratories, November (1998)
  8. S.B.Kim, et al., 'Evaluation of the Containment Integrity due to Direct Containment Heating in Korean 1300 MWe PWR,' Korea Atomic Energy Reseach Institute, Transactions of the 15th International Conference on Structural Mechanics in Reactor Technology (SMiRT-15), section XI, pp. 197-205, Seoul, Korea, August 15-20, (1999)
  9. K.K.Murata, et al., 'Code Manual for CONTAIN 1.2: A Computer Code for Nuclear Reactor Containment Analysis,' (Draft), Sandia National Laboratories, April, (1993)
  10. K.K.Murata, et al., 'Code Manual for CONTAIN 2.0: A Computer Code for Nuclear Reactor Containment Analysis,' NUREG/CR-6533, SAND97-1735, Rev. 0, Sandia National Laboratories, June, (1997)
  11. J.W.Park, et al., 'Estimation of Direct Containment Heating Loads in KNGR Using CONTAIN Code,' Korea Power Engineering Company, Inc., Presented at the 1999 KNS Spring Meeting, (CD-ROM), Pohang, Korea, May, (1999)
  12. Henry, R. E. et al., 'MAAP4-Modular Accident Analysis program for LWR Power Plants,' EPRI, May (1994)
  13. R.L.Iman and M.J.Shortencarier, ' A FORTRAN 77Program and User's Guide for the Generation of Latin Hypercube and Random Sample for Use with Computer Models,' NUREG/CR-3624, SAN83-2365, Sandia National Labortories, March, (1984)
  14. T.Y.Chu, et al., 'An Assessment of the Effects of Heat Flux Distribution and Penetration on the Creep Rupture of a Reactor Vessel Lower Head,' Twelfth Proceedings of Nuclear Thermal Hydraulics, 1997 ANS Annual Winter Meeting, pp. 135-144, Albuquerque, November, (1997)
  15. KEPCO, 'Ulchin Unit 3&4 Final Probabilistic Safety Assessment Report,' vol. 5
  16. D.C.Williams, et al., 'CONTAIN Code Analysis of Direct Containment Heating (DCH) Experiments: Model Assessment and Phenomenological Interpretation,' Proceedings of the ANS Winter Meeting, Thermal Hydraulics Division, pp. 495-497, San Francisco, October, (1995)